NUREG/IA-0434 

International 

 Agreement Report

The Development and Application of TRACE/PARCS 

Model for Lungmen ABWR 

Prepared by:

Jong-Rong Wang, Chia-Ying Chang*, Hao-Tzu Lin, Chunkuan Shih*

Institute of Nuclear Energy Research, Atomic Energy Council, R.O.C.

1000, Wenhua Rd., Chiaan Village, Lungtan, Taoyuan, 325, Taiwan

 

*Institute of Nuclear Engineering and Science, National Tsing Hua University

101 Section 2, Kuang Fu Rd., HsinChu, Taiwan

K. Tien, NRC Project Manager

 Division of Systems Analysis

Office of Nuclear Regulatory Research

U.S. Nuclear Regulatory Commission

Washington, DC 20555-0001

Manuscript Completed: April 2013

Date Published: December 2013

Prepared as part of 

The Agreement on Research Participation and Technical Exchange

Under the Thermal-Hydraulic Code Applications and Maintenance Program (CAMP)

Published by

U.S. Nuclear Regulatory Commission
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infringe privately owned rights. 
NUREG/IA-0434 

International 

 Agreement Report

The Development and Application of TRACE/PARCS 

Model for Lungmen ABWR 

Prepared by:

Jong-Rong Wang, Chia-Ying Chang*, Hao-Tzu Lin, Chunkuan Shih*

Institute of Nuclear Energy Research, Atomic Energy Council, R.O.C.

1000, Wenhua Rd., Chiaan Village, Lungtan, Taoyuan, 325, Taiwan

 

*Institute of Nuclear Engineering and Science, National Tsing Hua University

101 Section 2, Kuang Fu Rd., HsinChu, Taiwan

K. Tien, NRC Project Manager

 Division of Systems Analysis

Office of Nuclear Regulatory Research

U.S. Nuclear Regulatory Commission

Washington, DC 20555-0001

Manuscript Completed: April 2013

Date Published: December 2013

Prepared as part of 

The Agreement on Research Participation and Technical Exchange

Under the Thermal-Hydraulic Code Applications and Maintenance Program (CAMP)

Published by

U.S. Nuclear Regulatory Commission

iii

ABSTRACT

TRACE code developed by US NRC is an advanced thermal hydraulic system code for nuclear 

power plant safety analysis. The feature of 3-D reactor vessel simulation in TRACE allows 

complex flow modeling and this kind of vessel is employed in Lungmen TRACE model. PARCS 

is a multi-dimensional neutronic simulator which is able to incorporate a 3-D reactor core with 

one-to-one fuel assembly simulation for the realistic representative. The state-of-the-art is that 

PARCS is capable of being coupled to TRACE, which provides the temperature and flow field 

information to PARCS during calculations. In this research, the TRACE/PARCS model of 

Lungmen nuclear power plant is established first. Then, the TRACE/PARCS model is used for 

two transient analyses - 3 RIPs trip and HPCF inadvertent startup. These events could test the 

ability of the coupling of TRACE and PARCS. Moreover, SNAP, a user-friendly interface 

developed by the US NRC and APT, provides users a graphical working environment and 

enables to animate the simulation as an alternative presentation of calculation results.
iv

 
v

FOREWORD

The US NRC (United States Nuclear Regulatory Commission) is developing an advanced 

thermal hydraulic code named TRACE for nuclear power plant safety analysis. The development 

of TRACE is based on TRAC, integrating RELAP5 and other programs. NRC has determined 

that in the future, TRACE will be the main code used in thermal hydraulic safety analysis, and no 

further development of other thermal hydraulic codes such as RELAP5 and TRAC will be 

continued. A graphic user interface program, SNAP (Symbolic Nuclear Analysis Program) which 

processes inputs and outputs for TRACE is also under development. One of the features of 

TRACE is its capacity to model the reactor vessel with 3-D geometry. It can support a more 

accurate and detailed safety analysis of nuclear power plants. TRACE has a greater simulation 

capability than the other old codes, especially for events like LOCA. 

Taiwan and the United States have signed an agreement on CAMP (Code Applications and 

Maintenance Program) which includes the development and maintenance of TRACE. INER 

(Institute of Nuclear Energy Research, Atomic Energy Council, R.O.C.) is the organization in 

Taiwan responsible for the application of TRACE in thermal hydraulic safety analysis, for 

recording user’s experiences of it, and providing suggestions for its development. In this report, 

the TRACE/PARCS model of Lungmen NPP has been built and the analysis of 3 RIPs trip and 

HPCF inadvertent startup are performed. 
vi

 
vii

CONTENTS

Page

ABSTRACT ..................................................................................................................................... iii 

FOREWORD .................................................................................................................................... v 

CONTENTS ................................................................................................................................... vii 

FIGURES ........................................................................................................................................ ix 

EXECUTIVE SUMMARY ................................................................................................................ xi 

ABBREVIATIONS ........................................................................................................................ xiii 

1. INTRODUCTION.................................................................................................................... 1-1 

2. DESCRIPTION OF LUNGMEN MODELS ............................................................................ 2-1 

2.1 Lungmen TRACE Model ............................................................................................. 2-1 

2.2 Lungmen PARCS Model ............................................................................................. 2-2 

2.3 Lungmen TRACE/PARCS Coupling .......................................................................... 2-3 

3. RESULTS AND DISCUSSION .............................................................................................. 3-1

3.1 Steady State Results ................................................................................................... 3-1 

3.2 Transients Results ...................................................................................................... 3-3 

3.2.1 3 RIPs Trip ...................................................................................................3-3

3.2.2 HPCF Inadvertent Startup ..........................................................................3-5

4. CONCLUSIONS..................................................................................................................... 4-1 

5. REFERENCES ....................................................................................................................... 5-1 
viii

 
ix

FIGURES

Page

1. Lungmen TRACE model ................................................................................................ 2-1 

2. 206 CHAN assignment in Lungmen TRACE model .................................................... 2-2 

3. 205 control rod pattern of Lungmen PARCS model ................................................... 2-3 

4. Axial relative power difference between PARCS and SIMULATE-3 ......................... 3-2 

5. Radial relative power difference between PARCS and SIMULATE-3 ....................... 3-2 

6. The comparisons of core inlet flow for 3 RIPs trip .................................................... 3-3 

7. The comparisons of reactivity for 3 RIPs trip (a) Void reactivity, (b) Doppler 

reactivity, and (c) Total reactivity ................................................................................. 3-4 

8. The comparisons of power for 3 RIPs trip .................................................................. 3-5 

9. Responses of power, steam flow rate, core inlet flow rate and feedwater flow rate 

for HPCF inadvertent startup ........................................................................................ 3-6 

10. Responses of Water level and pressure rise for HPCF inadvertent startup............ 3-7 

11. Reactivity responses of the prediction analysis for HPCF inadvertent startup ...... 3-7 

12. Lungmen animation model for HPCF inadvertent startup ......................................... 3-8 
x

 

 

 
xi

EXECUTIVE SUMMARY

Lungmen nuclear power plant (NPP) is the first ABWR plant in Taiwan and still under 

construction. It has two identical units with 3,926 MWt rated thermal power each and 52.2×106 

kg/hr rated core flow. The core has 872 bundles of GE14 fuel, and the steam flow is 7.637×106

kg/hr at rated power. There are 10 RIPs (reactor internal pumps) in the reactor vessel, providing 

111% rated core flow at the nominal operating speed of 151.84 rad/sec. 

According to the user manual, TRACE is the product of a long term effort to combine the 

capabilities of the NRC’s four main systems codes (TRAC-P, TRAC-B, RELAP5 and RAMONA) 

into one modernized computational tool. NRC has ensured that TRACE will be the main code 

used in thermal hydraulic safety analysis in the future without further development of other 

thermal hydraulic codes, such as RELAP5 and TRAC. Besides, the 3-D geometry model of 

reactor vessel, which is one of the representative features of TRACE, can support a more 

accurate and detailed safety analysis of NPPs. On the whole TRACE provides greater simulation 

capability than the previous codes, especially for events like LOCA. 

PARCS is a multi-dimensional reactor core simulator which involves a 3-D calculation model for 

the realistic representation of the physical reactor while 1-D modeling features are also available. 

PARCS is capable of coupling the thermal-hydraulics system codes such as TRACE directly, 

which provide the temperature and flow field data for PARCS during the calculations. 

This research focuses on the development of the Lungmen NPP TRACE/PARCS model. The 

TRACE/PARCS model is used for two transient analyses - 3 RIPs trip and HPCF inadvertent 

startup. These events could test the ability of the coupling of TRACE and PARCS. The trends of 

these parameters are generally in consistent with FSAR data. Besides, the animation model of 

Lungmen NPP is also established by using SNAP code. The analysis results of TRACE/PARCS 

can be observed in the animation model of Lungmen NPP. 
xii

 
xiii

ABBREVIATIONS

ABWR Advanced Boiling Water Reactor 

APT Applied Programming Technology 

CAMP Code Applications and Maintenance Program 

ECCS Emergency Core Cooling System 

FSAR Final Safety Analysis Report 

HPCF High Pressure Core Flooder 

HPCI High Pressure Coolant Injection 

INER Institute of Nuclear Energy Research 

NRC Nuclear Regulatory Commission 

NPP Nuclear Power Plant 

PARCS Purdue Advance Reactor Core Simulator 

RCIS Rod Control & Information System 

RCPB Reactor Coolant Pressure Boundary 

RFCS Recirculation Flow Control System 

RPV Reactor Pressure Vessel 

SCRRI Selected Control Rod Run-In 

SNAP Symbolic Nuclear Analysis Package 

TPC Taiwan Power Company 

TRACE TRAC/RELAP Advanced Computational Engine 
xiv

 
1-1 

1. INTRODUCTION

Lungmen NPP, a GE-designed ABWR, is owned by Taipower Company. It has two identical 

units with 3926 MWt rated core thermal power and 52.2×106 kg/hr rated core flow each. The 

reactor core is comprised of 872 fuel assemblies and will be loaded with GE14 fuel in the initial 

core. Each GE14 fuel assembly has 10×10 arrays consisting of 92 fuel rods and 2 water rods 

with each occupying four equivalent fuel rod positions. The annular region in the lower 

downcomer of the reactor pressure vessel accommodates 10 RIPs which can provide core flow 

up to 111% of the rated capacity. Six of the 10 RIPs are connected to a motor-generator set 

(M-G Set) while the remaining 4 RIPs are not connected as such and are powered directly by the 

13.8 kV Bus. 

TRACE code developed by NRC is an advanced thermal hydraulic system code for nuclear 

power plant safety analysis. The feature of 3-D reactor vessel simulation in TRACE allows 

complex flow modelling and this kind of vessel is employed in Lungmen TRACE model. One of 

the features of TRACE is 3-D reactor vessel simulation which allows complex flow modeling. 

PARCS is a multi-dimensional neutronic simulator which is able to incorporate a 3-D reactor 

core with one-to-one fuel assembly simulation for the realistic representative. The 

state-of-the-art is that PARCS is capable of being coupled to TRACE, which provides the 

temperature and flow field information to PARCS during calculations. TRACE is responsible for 

NPP safety analysis simulating thermal-hydraulic system, and PARCS is responsible for the 

neutronic calculation in core according to the variations evaluated from TRACE. TRACE/PARCS 

model are believed to be the best-estimate calculations. 

This research focuses on the development of the Lungmen NPP TRACE/PARCS model. 

Lungmen NPP TRACE/PARCS model is used for two transient analyses - 3 RIPs trip and HPCF 

inadvertent startup. These events could test the ability of the coupling of TRACE and PARCS. 

The transient data from FSAR [1] have been used to compare the analysis results of Lungmen 

NPP TRACE/PARCS model. Moreover, SNAP, a user-friendly interface developed by the U.S. 

NRC and APT, provides users a graphical working environment and enables to animate the 

simulation as an alternative presentation of calculation results. 
1-2 

 
2-1

2. DESCRIPTION OF LUNGMEN MODELS

2.1 Lungmen TRACE Model

SNAP v 2.2.1 and TRACE v 5.0p3 are used in this research. First, the system and operating 

data [1]-[5] for Lungmen NPP are collected. Second, several important control systems such as 

RIPs control system, steam bypass and pressure control system and feedwater control system 

etc. are established by SNAP and TRACE. Third, other necessary components (e.g., reactor 

pressure vessel and main steam piping) are added into the TRACE model to complete the 

TRACE model for Lungmen NPP. Figure 1 shows the major thermal-hydraulic components 

simulated in Lungmen NPP TRACE model. The 3-D VESSEL component is comprised of 11 

axial levels, 4 radial rings where the outer ring models the downcomer, and 6 azimuthal sectors 

(separately in 36°, 36°, 108°, 36°, 36°, 108° apart) for a total of 264 computational cells. The 4 

separate main steam lines are connected to the four 36° sectors, the 6 feedwater lines are 

connected to each azimuthal sector individually, and the 2 loops of HPCF system are connected 

to the 6th axial level of VESSEL above the core as design. Ten RIPs are arranged into three 

groups. The first two groups, with 3 RIPs each, are connected to M-G sets. The third group, 

which has 4 RIPs, is not connected to M-G sets. The simulation for core in Lungmen NPP 

TRACE model has been modified by replacing 206 CHAN components for the previous 18 

CHAN components for finer calculation in core [3]. In TRACE model, the CHAN component is in 

lieu of fuel assemblies in core; the 872 fuel assemblies are assigned to 206 CHAN components 

with each CHAN component consisting of 3 to 7 fuel assemblies and the assignment is shown in 

Figure 2; this assignment is also mapped to Lungmen NPP PARCS model for coupling. 

Figure 1 Lungmen TRACE model
2-2

Figure 2 206 CHAN assignment in Lungmen TRACE model

2.2 Lungmen PARCS Model

In this work, 3-D core simulator PARCS V 3.0 is used. The model composed of 872 neutronic 

nodes representing 872 fuel assemblies one-to-one in radial direction. The active core height is 

381 cm, which is modeled in PARCS with 25 axial layers. At the top and bottom of the active 

core, there exist two 15.24-cm-thick axial reflector regions. A total 205 control rods are also 

simulated as shown in Figure 3. 

When processing, the PARCS model relies on the use of a special macroscopic cross-section 

file, PMAX. GenPAMXS code reformats the outputs of the lattice physics code, CASMO-4 in this 

work, as PMAX files which is able to comply with the cross-section format required in PARCS. 
2-3

Figure 3 205 control rod pattern of Lungmen PARCS model 

2.3 Lungmen TRACE/PARCS Coupling

In TRACE/PARCS model, the overall controls of the coupled transient, such as convergence 

checks and trip initiation are handled by TRACE. For fast steady state initialization, a neutronic 

calculation skipping strategy is used, i.e., PARCS calculation was done only once every 10 time 

advances in TRACE. In addition, the reflector nodes in PARCS are not mapped to 

thermal-hydraulic channels since the reflector thermal-hydraulic properties are fixed. Fuel 

assemblies are mapped into 206 thermal-hydraulic CHANs as shown in Figure 2 according to 

the distribution of the CHANs in TRACE model. The numbers in the figure indicate channel 

number assignments of the fuel assemblies and “0” corresponds to the reflector region. 
2-4

 
3-1

3. RESULTS AND DISCUSSION 

3.1 Steady State Results 

To ensure that the data at steady state in TRACE model for Lungmen nuclear power plant are 

consistent with FSAR chapter 15, the steady state analysis should be conducted prior to 

transient analysis. As shown in Table 1, the steady state results are in good agreements with 

FSAR. 

Table 1 The comparisons of FSAR and TRACE/PARCS for the steady state 

Parameter FSAR TRACE standalone / 

Error (%) 

TRACE/PARCS / 

Error (%) 

Power (MWt) 3926 3926 / 0 3926 / 0 

Steam flow (kg/s) 2122 2121.2 /0.04 2121.2 / 0.04 

Core flow (kg/s) 14500 14487 / 0.09 14610 / 0.76 

Dome pressure (MPa) 7.2 7.16 / 0.56 7.16 / 0.56 

NRWL (m) 13.42 13.43 / 0.07 13.43 / 0.07 

Lungmen SIMULATE-3 model, a verified core physics code, was used to examine the 

performances of Lungmen TRACE/PARCS model. The results of comparing the differences 

between TRACE/PARCS and SIMULATE-3 in axial and radial directions are shown as Figure 4 

and Figure 5, respectively. In Figure 4, the axial relative power profile calculated by 

TRACE/PARCS is compared against SIMULATE-3 at the same condition. The distributions of 

the axial relative power profiles of both codes are similar, especially at the upper part of the core.

As shown in Figure 5, the maximum difference in radial direction is under 25% and the 

difference is symmetric. 
3-2

Figure 4 Axial relative power difference between PARCS and SIMULATE-3 

Figure 5 Radial relative power difference between PARCS and SIMULATE-3 
3-3

3.2 Transient Results 

3.2.1 3 RIPs Trip 

In 3 RIPs trip transient, the core inlet flow drops immediately following the trip of RIPs. The 

reduction of core inlet flow leads to the increase of core void fraction and also negative reactivity 

which results in the decrease of the core power. At about 6 second, the core flow and power 

reaches its steady state. The simulation results of selected important system parameters are 

compared against FSAR data. 

The trip of 3 RIPs causes the core inlet flow to decrease simultaneously as shown in Figure 6, 

and the core inlet flow become steady at around 1.5 second. The amount of void in core 

becomes greater due to the drop of the core inlet flow. Figure 7 indicates that the significant void 

reactivity feedback dominants the total reactivity. As a result, the core power drops 

corresponding to reactivity variations as presented in Figure 8. The trends of these parameters 

are generally in consistent with FSAR data. 

0 1 2 3 4 5 6 7 8 9 10

Time (s)

60

70

80

90

100

C

ore inlet flo

w (%)

3 RIPs trip-Core inlet flow

FSAR

TRACE/PARCS

Figure 6 The comparisons of core inlet flow for 3 RIPs trip 
3-4

0 1 2 3 4 5 6 7 8 9 10

Time (s)

-0.8

-0.6

-0.4

-0.2

0

0.2

0.4

Reactivity ($)

3 RIPs trip-Void Reactivity

FSAR

TRACE/PARCS

0 1 2 3 4 5 6 7 8 9 10

Time (s)

-0.8

-0.6

-0.4

-0.2

0

0.2

0.4

Reactivity ($)

3 RIPs trip-Doppler Reactivity

FSAR

TRACE/PARCS

(a) (b)

0 1 2 3 4 5 6 7 8 9 10

Time (s)

-0.8

-0.6

-0.4

-0.2

0

0.2

0.4

Reactivity ($)

3 RIPs trip-Void Reactivity

FSAR

TRACE/PARCS

(c)

Figure 7 The comparisons of reactivity for 3 RIPs trip 

(a) Void reactivity, (b) Doppler reactivity, and (c) Total reactivity 
3-5

0 1 2 3 4 5 6 7 8 9 10

Time (s)

50

60

70

80

90

100

P

o

w

er (%)

3 RIPs trip-Power

FSAR

TRACE/PARCS

Figure 8 The comparisons of power for 3 RIPs trip 

3.2.2 HPCF Inadvertent Startup 

In Lungmen NPP, inadvertent startup of the HPCF System is the “increase in reactor coolant 

Inventory” transient, supposedly due to system malfunction or operator error. HPCF is comprised 

of two loops flooding water to the RPV above the core. HPCF inadvertent startup is postulated 

that under normal operation, one loop of the HPCF introduces cold water of 40°C into the upper 

core plenum, and the full HPCF flow is established within 1 second. The addition of cooler water 

to the upper plenum results in a reduction in steam flow, causing slight depressurization in the 

RPV. Despite the transient is mild, the asymmetric injection in the RPV will make temperature 

and flow field distribute unevenly. 
3-6

Figure 9 and 10 present the system response for the simulated transient event, inadvertent 

startup of HPCF. One loop of the HPCF system begins to introduce cold water over the core in 

the RPV, and within 1 second, the full HPCF flow is established. The additional cooler water into 

the upper plenum leads to a reduction in steam flow and causes slight depressurization as the 

pressure regulator responds to the event. In Figure 11, the reactivity responses are no 

significance. Therefore, in that thermal and pressure variations are relatively small. Hence, this 

event is a mild transient. The HPCI system in traditional BWRs injects coolant through the 

downcomer together with feedwater into the reactor core [5]. Compared with the BWRs, the 

introduction of HPCF over the core in Lungmen ABWR affects the core much more directly. 

Even though the inadvertent startup of HPCF is a mild event for Lungmen ABWR, the 

asymmetry injection in the RPV would make temperature and flow field distribute unevenly. The 

3-D vessel simulation in TRACE together with 3-D core simulation in PARCS enables the 

evaluation for local feedback effects and local responses such that the animation under SNAP 

interface is able to demonstrate the asymmetric phenomenon of the distribution of temperature, 

fluid condition, and relative power with 3-D display. The animation model is to embody the 

calculation results with not only coloured plant icons but also strip plots of the data channel. 

Figure 12 shows a set of successive screen shots of the animation. As shown in the figure, it’s 

obvious to see the injection of only one loop of HPCF over the core; the strip plot at the bottom 

presents power, steam flow rate, and feedwater flow rate in percentage with respect to time; the 

distributions of core power, moderator temperature, and fuel temperature shown on the right 

hand side; the status of each layer of the VESSEL including void fraction and fluid condition are 

presented on the left hand side. Through the demonstration of Figure 12, the uneven distribution 

in the RPV as well as several variations can be observed simultaneously. 

0 1 2 3 4 5 6 7 8 9 10

Time (s)

50

60

70

80

90

100

110

120

P

ercent of rated

Power

Steam flow rate

Core inlet flow rate

Feedwater flow rate

Power (FSAR)

Steam flow rate (FSAR)

Core inlet flow rate (FSAR)

Feedwater flow rate (FSAR)

Figure 9 Responses of power, steam flow rate, core inlet flow rate and feedwater flow 

rate for HPCF inadvertent startup 
3-7

0 1 2 3 4 5 6 7 8 9 10

Time (s)

0.4

0.8

1.2

1.6

2

Level fro

m dry skirt (m)

NRWL(m)

Dome pressure rise (MPa)

Level 3(m)

NRWL(m) (FSAR)

Dome pressure rise (MPa) (FSAR)

Level 8 (m)

-0.2

-0.1

0

0.1

0.2

Pressure (M

Pa)

Figure 10 Responses of Water level and pressure rise for HPCF inadvertent startup 

0 1 2 3 4 5 6 7 8 9 10

Time (s)

-0.5

-0.4

-0.3

-0.2

-0.1

0

0.1

0.2

0.3

0.4

0.5

R

e

activity ($)

Void reactivity

Doppler reactivity

Control rod reactivity

Total reactivity

Void reactivity (FSAR)

Doppler reactivity (FSAR)

Control rod reactivity (FSAR)

Total reactivity (FSAR)

Figure 11 Reactivity responses of the prediction analysis for HPCF inadvertent startup 
3-8

Figure 12 Lungmen animation model for HPCF inadvertent startup 
4-1

4. CONCLUSIONS

By using SNAP/TRACE/PARCS, this study has developed the TRACE/PARCS model of the 

Lungmen NPP. There are 206 CHAN components, representing 872 fuel assemblies in core, in 

order to achieve finer simulation in core. The 3-D VESSEL component in TRACE model 

incorporating 3-D core simulation by PARCS enables the evaluations of local feedback effects 

as well as local responses. The proposed TRACE/PARCS model analyzed the transients of 3 

RIPs trip and HPCF inadvertent startup. Analytical results indicate that the Lungmen NPP 

TRACE/PARCS model can predict the behaviors of important plant parameters reflecting 

consistent trends with FSAR data. As a result, the Lungmen 3-D animation under the interface of 

SNAP is adequate for the asymmetric phenomenon due to only one loop of HPCF system 

injection, in which one could observe the variations in the Lungmen ABWR during the transient. 
4-2

 
5-1 

5. REFERENCES

1. Taiwan Power Company, Final Safety Analysis Report for Lungmen Nuclear Power Station 

Units 1&2 (FSAR), 2007. 

2. J.R. Wang, H.T. Lin, Y.H. Cheng, W.C. Wang, C. Shih, “TRACE models and verifications for 

Lungmen ABWR”, American Nuclear Society Winter Meeting, U.S.A., November 15-19, 

2009. 

3. H.T. Lin, J.R. Wang, C. Shih, The development and assessment of TRACE model for 

Lungmen ABWR, KERNTECHNIK, Vol.76, 2011. 

4. Taiwan Power Company, Lungmen ABWR Training Manual, 2006. 

5. Y.T. Ku, Y.S. Tseng, C. Shih, C.W. Su, J. R. Wang, “The Thermal Hydraulic Behavior in 

Core Inlet Region of BWR with the Inadvertent Startup of HPCI”, The 9th International 

Topical Meeting on Nuclear Thermal-Hydraulics, Operation and Safety (NUTHOS-9), 

Kaohsiung, Taiwan, September 9-13, 2012. 






UNITED STATES 

NUCLEAR REGULATORY COMMISSION

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--------------------

OFFICIAL BUSINESS
NUREG/IA

-0434 The Development and Application of TRACE/PARCS Model 

for Lungmen ABWR